Arsenate carrier precipitation method of separating plutonium from neutron irridiated uranium and radioactive fission products



-ried out in a neutronic reactor of the pile type. -a reactor neutrons react with U to form U which -rays.

gerous to personnel and the material must be handled;

2,989,367 I ARSENATE CARRIER PRECIPITATION METHOD OF SEPARA-IING PLUTONIUM FROM NEUTRON IRRIDIA'I'ED URANIUM AND RADIOACTIVE FISSION PRODUCTS Stanley G. Thompson, Daniel R. Miller, and Ralph A. James, Richmond, Califi, assignors to the United States of-America as represented by the United States Atomic Energy Commission No Drawing. Filed Aug. 11, 1947, Ser. No. 768,064 7 Claims. (Cl. 23-145) decays with a half-life of 23 minutes to form Np which in turn decays with a half-life of 2.3 days to form plutonium. The uranium usually used in neutronic reactors is a mixture of U and U and during the reaction neutrons also react with the U to formv fission products. These consist mainly of elements with atomic numbers products are usually highly radioactive and decay to more stable elements by the emission of beta and gamma The radioactivity of these elements is very danwith extreme care.-

The uranium mass as it is removed from the neutronic reactor usually contains a very large percentage of unreacted uranium and a very small percentage of neptunium, plutonium and radioactive fission products, and,

in order to obtain plutonium in a usable form, it is necessary to separate plutonium from the unreacted uranium and radioactive fission products.

The uranium mass is commonly permitted to age for a short period in order that the neptunium contained therein may decay to plutoniumsince neptunium has a half-life of only 2.3 days. However, theseparation of plutonium from the other contaminating elements of the mass requires very careful processing. This processing may be arbitrarily divided into three steps. 'The' first Un t Sta e Patent from approximately to 45 and 51 to 60. These fission step is usually the separation of the .plutonium from the uranium and certain of the fission products. This step is usually referred to as extraction. In the second step the plutonium is separated from the remaining radioactive fission products and this step is usually referred to as decontamination. The final step is the plutonium concentration step.

Several methods of separating plutonium from conj taminating elements have been worked out. 'The one 2 most widely used at present is a method based upon the fact that plutonium has several oxidation states ,and that certain plutonium compounds in the various states have different solubilities. Where plutonium is present in sufiicient concentration in the mass of neutron irradiated uranium, the mass may be dissolved and an insoluble precipitate of plutonium may be formed which will remove plutonium from certain dissolved contaminants.

While plutonium may be present in solution in a concentration great enough that it may be directly precipitated from the solution, plutonium is usually present in a neutron irradiated mass of uranium in such small quantity that the plutonium cannot be directly precipitated from a solution formed from the irradiated mass. In this case plutonium may be separated by forming a carrier precipitate which will carry the plutonium from solution. This carrier precipitate for plutonium, usually referred to as the product precipitate, may then be redissolved, the oxidation state of plutonium changed and another precipitate formed in the solution which will not carry the plutonium in its new oxidation state but will carry contaminating elements contained in the solution. This precipitate is usually referred to as a by-product precipitate. In order to increase the efficiency of the carriers in separating plutonium from contaminating elements it is often necessary to add compounds to the solution which will prevent the undesirable elements from being carried with the element which it is desired to remove from solution, by increasing the solubility of the contaminating element. These compounds are referred to as complexing agents. Thus, in the precipitation of plutonium from uranium, sulfate ions may be added to the solution 'to form a more soluble uranium complex thus preventing the uranium from being carried with the plutonium carrier precipitate.

The principal object of this invention is to provide a novel and simplified method of extracting plutonium from neutron irradiated uranium.

Another object of this invention is to provide a carrier for plutonium in the +4 oxidation state which will not form insoluble compounds with the UO ion.

Other objects and advantages of this invention will be come apparent in the following detailed description.

We have discovered that plutonium in a +4 oxidation state forms a compound with the arsenate ion which, while soluble in a highly acid solution, is very insoluble in a neutral or basic solution. This compound, which has the formula P11 (AsO.;) may be formed by the addition of a soluble arsenate compound such as arsenic acid to a solution containing the plutonium ion in the +4 oxidation state followed by neutralization of said solution. This precipitate is a dense crystalline substance which may be easily separated from the supernatant liquid by any of the usual methods. The solubility of 3 N nitric acid. Separation of plutonium from contami- ;nants such as uranium and many of the fission products, particularly the radioactive fission products which do not 0 form arsensates insoluble in aqueous solution, may be readily effected by simply adding an excess of arsenate ion to an acid solution containing said plutonium, uranium and fission products in an ionized state.

In most neutron irradiated masses containing plutonium the plutonium is present in concentrations so low that the plutonium cannot be directly precipitated from the solution. In such cases it is necessary to use carrier techniques to separate the plutonium. It is desirable to have a carrier precipitate that will be able to carry plutonium from solution quantitatively without precipitating or carrying other undsirable elements contained in solution. In addition, it is desirable that a carrier carry plutonium quantitatively at all concentrations from tracer to the concentration at which plutonium may be directly precipi- One phase of this invention is concerned with the separation of plutonium from a solution formed by dissolving in an inorganic acid a mass of uranium which has been irradiated with neutrons in a neutronic reactor employing a chain reaction. Plutonium ions are present in said solution in a valence state not greater than +4 and in a concentration greater than tracer but less than that from which plutonium compound could be directly precipiated. The neutron irradiated uranium mass also contains small quantities of fission products with the concentration approximately equal to that of the plutonium. The con centration of the uranyl nitrate hexahydrate has been adjusted in the solution so that it is below about 22% in order that the loss of element 94 in the extraction process tated and that the conditions of acidity at which the prehe p to a minimumcipitate forms be not so rigid that it is impractical to use Y the pl'oeeSS of this ihvehfioh a h' of arsehate the carrier on large scale operations. We have discovered P b y e i e is addefi e 50111391 that a carrier of bismuth arsenate fulfills all of these re- The eeldlty 0f the Sohlhen S then J e i dependlhg quirements and is a particularly suitable carrier for the uP011 the e h and Whether a eemplexlhg agent for gxtragtion of plutonium from a solution of neutron irrathe i gi y b 10I1 h p l d P se g of iate uranium. a so u e ismut compoun a'precipitate o ismut ar- The solubility of bismuth arsenate has been found to Senate Will form which will carry the Plutonium l'ehepfesbe approximately 20 mg/I, in 3 N HNO I 10 N HNQ3 ent in solution but not the uranyl ions. Although a preits solubility is about half that of bismuth phosphate. The 2 eiPitete y he jifel'medy w Solution at room precipitacie is dense and crystalline and thus may be readily 5 P f dg e ghp' 'g g to heat the 31 separate from solution. prior to t e a ition o e ismut ions to a mo erate Bismuth arsenate has been found to be a very efficient temperature of for example 75 ahd Somewhat better lu+ carrier. It is believed that the principal reason for results f been obtamed by dlgestlng the Preclpnate its suitability is that the bismuth arsenate precipitate is i P i f mgjml' isomorphous with plutonium arsenate. Thus, it is be- 0 ,lsmu loin as o e a llieved that plutonium ions are incorporated in the crystal i g gg gffig ii g i gpr g f :2: 21 2 2 attice structure of bismuth arsenate when that 'com- 1 P e $6 In "i gaunt! is formed in the solution containing the Pu+ ion. a: i ggigjiifhZ- ltgfggwizfi t igl g gl: owever, some of the carrying eifect may also be due to 35 v e m a Surface adsorption of the plutonium ion by the bismuth solubleb smuth salts, for example the sulfate or halides arsenate crystal. Bismuth arsenate has been found to ig n i d 1 h carry plutonium quantitat vely at a wide range of'concenwg z y P 5. t f if f s i trations. Thus, the carrying of plutonium is effective not iiif h ls a n y n i m only at tracer concentrations but at all'concentrations up 40 3 g f g g i fg :3 1 3222; 2:8 a $3533; to the concentration of In i I I precipitated with the g fg g may be dlrecfly of nitric and sulfuric acids has been found to be particu- One aspect of carrying techniques) in which bismuth iii: suitable for the formation of the arsenate preciparsenate i prfwed ltseif to be greatly superior to most One embodiment of this invention is concerned with i i carriers is h range o concentrations and the separation of plutonium from uranium without the :g a at g i a i l? g gf precipitate i us of a complexing agent for the uranyl ion. In most e ilency W1 nlsmuth arsenate W111 other processes for the "separation of plutonium from P utomflm t P i range- Many f P uranium the difference in solubility between the plutonium toniurii earner precipitates will c y plutonium ions salt formed with the carrier and'the uranyl salt formed q f f y y when the ph h 1S p h is so small that without the use of a complexing agent a i n w in n rr w yoncentration limi r w 1l f i to substantial percentage :of'the uranyl ion is carried with the separate plutonium 10118. from other contaminating ions, product carrier precipitate. In order to avoid forming a unless the precipitation is carried out within very narrow uranyl compound which will be carried with the pluacidity concentration limits. Numerous experiments have tonium carrier precipitate, a complexing agent is used to been carried out to determine the carrying power of bisincrease the solubility of the uranyl ion and this is commuth arsenate for plutonium over a wide range of conmonly a soluble sulfate compound. Plutonium arsenate, centrations and reaction conditions. The conditions and h wever, is more ins lu le nd r ni m r n m r results of some of these experiments are summarized in soluble than the corresponding compounds for .most other Table I following: carriers so that plutonium may be carried from solut on TABLE I Carrying of plutonium by BiAsO Digestion Exp. Percent HNO Bl 'HSO ditl P t No. U02(NO3)i-6H20 cone, As0 00110., mine, .00 on! P u gl lg- N eonc., mg./ N clpitated M cc. time 0.

(hrs) '3 0.10 2.5 s5 an 99.0 0.2 0.50 2.5 1 '2 99.1 0.2 0. 50 2.5 1 2 75 09.4 0.2 0.50 2.5 2 76 99.8 0. 2 0. 50 2. 5 2 75 96.8 1 0.10 1.25 M 75 96.3

without a uranyl ion complexing agent may be illustrated by the following example.

EXAMPLE I A mass of neutron irradiated uranium was dissolved in a nitric acid solution and the concentration of the uranyl nitrate hexahydrate was adjusted to of thesolution. The nitric acid concentration was then adjusted to l N. Arsenic acid was added'to make the solution 0.1 M and the solution heated to 75 C. A quantity of Bi(NO sufiicient to give a concentration of 1.5 mg. of Bi+ per ml. of solution was added to the solution and the solution digested for two hours at 75 C. with mechanical agitation. The solution was then filtered to remove the carrier precipitate. Tests of the carrier precipitate showed that 99.3% of the plutonium had been extracted. This experiment was repeated numerous times with the nitric acid concentration being varied between 0.5 and 1.0 N and this sequence of experiments disclosed that this process worked equally well throughout this range of concentrations with at least 98% of the plutonium being carried in all cases.

Where a complexing agent such as sulfate ion is not used in the extraction phase it has been found that the conditions for achieving maximum efficiency for the extraction step are as follows: If the H AsO concentration is 0.1 M no uranyl arsenate precipitates, if the HNO concentration is about 0.5 N. If the solution is l N in HNO the H AsO concentration should be maintained below 0.15 M to avoid uranyl arsenate precipitation.

. Although comparatively satisfactory carrying of plutonium may be achieved by the process of this invention without a complexing agent for the uranyl ion, it is also possible and in many cases desirable to use a uranyl ion complexing agent for the extraction of plutonium since a still wider range of acid concentrations may then be used. The normal complexing agent used is the sulfate ion and this may be added to the solution either as a soluble sulfate salt or by the use of sulfuric acid to adjust the acidity of the uranyl nitrate hexahydrate solution. It has been found that if the H AsO concentration is 0.4 M and the total additional acid concentration s-lz 4) is l N the H 80 concentration should be equal to or greater than 0.5 N. The use of the process of this invention for the extraction of plutonium with a complexing agent present may be illustrated by the following example.

EXAMPLE II Neutron irradiated uranium was dissolved in nitric acid to achieve a 20% uranyl nitrate hexahydrate solution. The nitric acid concentration was then adjusted to 0.2 N and sulfuric acid added to make the solution 1 N in total acid concentration. Arsenic acid was then added to the solution to .50 M concentration and the solution heated to 75 C. A quantity of Bi(NO suflicient to give a concentration of 2.5 mg. of Bi+ per ml. of solution was then added to the solution and the solution digested for two hours at 75 C. The bismuth arsenate precipitate was separated by centrifugation and it was found that 99.4% of the plutonium was carried with the precipitate.

During the extraction step plutonium is separated from uranium and from a majority of the fission products;

Ihowever, there are certain fission products which are isomorphous with plutonium in the +4 valence state and these fission products are usually carried with plutonium by the carrier precipitate. In order to separate plutonium from these fission products, a decontamination step is carried out whereby the plutonium is oxidized to the +6 oxidation state and a by-product carrier precipitate with these fission products formed and separated. Fission products which are most troublesome in this step are zirconium and niobium. In other processes the carrier precipitate often forms such insoluble compounds with the zirconium and niobium that when the carrier product is dissolved following the extraction step these fission product precipitates are not completely dissolved and their presence leads to product loss in the fission by-product precipitate and to low decontamination factors. Investigation has disclosed, however, that the arsenates of zirconium and niobium are sufliciently soluble that this difficulty is precluded in the process of this invention.

' The process of this invention may be used for the complete extraction and decontamination of plutonium. Thus, the plutonium may be extracted as described above and following this extraction the carrier precipitate may be dissolved in an inorganic acid, the plutonium oxidized to the +6 state and a by-product carrier of bismuth arsenate precipitated and separated from the solution. To complete the decontamination cycle, the Pu' ions may then be reduced and removed from solution by precipitating with a carrier precipitate of bismuth arsenate. This decontamination cycle may be repeated as often as necessary to obtain the desired concentration. In another embodiment, however, the bismuth arsenate extraction process may be combined with other methods of decontamination, for example with the bismuth phosphate decontamination process. The combination of the process of this invention and the bismuth phosphate decontamination process may be illustrated by the following example.

EXAMPLE III A neutron irradiated uranium mass was dissolved in nitric acid to produce a 20% solution of uranyl nitrate hexahydrate and the solution was then adjusted so that it contained 1 N HNO and 0.4 M H AsO A quantity of Bi(NO sufficient to give a concentration of 1.5 mg. of Bi per ml. of solution was added to the uranyl nitrate hexahydrate solution. The solution was then digested for two hours with mechanical agitation at C. to insure complete precipitation of the bismuch arsenate plutonium carrier. The plutonium carrier precipitate was separated by centrifugation, and a decontamination cycle carried out with the precipitate, using the standard bismuth phosphate method. This cycle comprised dissolving the product precipitate, that is the plutonium carrier precipitate of bismuth arsenate, in a nitric acid solution, oxidizing the plutonium ions to the +6 valence state with NaBiO forming a by-product carrier precipitate of BiPO in the solution, and separating this precipitate with accompanying fission products from the solution. The Pu+ ions were then reduced with ferrous ions, and the reduced plutonium ions precipitated from the solution with a carrier of BiPO Counter analysis showed that the product precipitate contained 98.3% of the plutonium present in the original solution, and that an overall decontamination factor of 2230 for ,8 rays, 164 for 'y rays was obtained.

In order to compare these results with those obtainable with the standard bismuth phosphate procedure for plutonium extraction and decontamination, an experiment was carried out with an equal quantity of uranyl nitrate hexahydrate containing the same amount of plutonium and fission products as that used in the preceding experiment, but using the standard bismuth phosphate procedure. The extraction cycle differed from that of the previous experiment in that the product carrier precipitate was bismuth phosphate rather than bismuth arsenate. The procedure followed in the decontamination cycles of the two experiments was the same. The advantages of using' the bismuth arsenate process are clearly shown by the results obtained, which are tabulated below.

. s tity to finish a maximum concentration of 0.4 Mg, contacting the solution with a soluble bismuth compound,

Experiment I Experiment II BlAsQ Extraction Oycle B1]? Decontamination, Cycle 1311 0 Extraction Cycle BiP O Decontamination Cycle Instead of using the bismuth phosphate oxidation-reduction procedure as described above for the decontamination cycles, other procedures may be used with the. process of this invention.

While there have been described certain embodiments of our invention it is to be understood that it is capable of many modifications. Changes therefore may be made without departing from the spirit and scope of this invention as described in the appended claims.

What is claimed is:

l. A method of separating plutonium from an aqueous solution which comprises precipitating the plutonium as plutonium arsenate.

2. In a process of separating plutonium from an aqueone solution, the step which comprises co-precipitating plutonium with bismuth arsenate.

3. The method of separating plutonium from uranium and radioactive fission products which are non-isomorphic with the plutonium ion in the +4 oxidation state, which comprises co-precipitating plutonium with bismuth arsenate.

4. The method of separating plutonium from uranium contained in an inorganic acid solution, which comprises co-pr'ecipit-ating plutonium with bismuth arsenate.

5. In a method of separating plutonium contained in an inorganic acid solution in an oxidation state of not greater than +4 from uranium, the step which comprises co-precipitating plutonium with bismuth arsenate.

6. The method of separating radioactive niobium values from hexavalent uranium values contained in an aqueous solution which comprises forming a 0.5-1.0 N nitric acid solution containing said niobium values and uranium values, adding orthoarsenic acid to said solution in a quanwhereby a bismuth arsenate carrier precipitate containing niobium is formed, and separating said precipitate from the solution.

7. The method of separating radioactive zirconium values from hexavalent uranium values contained in an aqueous solution which comprises forming a 0.5-1.0 N nitric acid solution containing said zirconium values and uranium values, adding orthoarsenic acid to said solution in a quantity to furnish a maximum concentration of 0.4 M, contacting the solution with a soluble bismuth compound, whereby a bismuth arsenate carrier precipitate containing zirconium is formed, and separating said precipitate from the solution.

References Cited in the file of this patent Mellor: Inorganic and Theoretical Chemistry, vol. 9, pp. 188 and 197 (1929). Published by Longmans, Green and Company, London (1929).

Hahn: Applied Radiochemistry, pp. -69 (1936). Cornell University Press, Ithaca, New York.

MUC-GTS2148 (N-2205 US. Atomic Energy Commission document dated January 16, 1946, declassified November 22', 1957; pages 1-3 and 38. This document disclosed information reported in CN-914 (September 1943) and from CN-l04l (October 1943) and these earlier dates are relied on.

Friend: Textbook of Inorganic Chemistry, Vol. VIII, part III (1926); Charles Griifin and Co., Ltd., London; pages 332-333.

Rodden: Analytical Chemistry of the Manhattan Project (NN'ES VIII-l), McGraw-Hill Book Co. (1950); page 22. 

1. A METHOD OF SEPARATING PLUTONIUM FROM AN AQUEOUS SOLUTION WHICH COMPRISES PRECIPITATING THE PLUTONIUM AS PLUTONIUM ARSENATE. 